1. Field of the Invention
The invention relates to methods for the treatment of radioactive waste products to facilitate their storage and disposal.
2. Description of the Prior Art
In nuclear energy production, the waste products generated in the reactor fuel elements are fission products which, for the most part, are radioactive. To a predominant degree, they belong to the medium heavy elements having atomic numbers of between 30 (zinc) and 64 (gadolinium). As long as the fuel elements are undamaged, these fission products remain enclosed in the sheathing tubes of the fuel elements. However, when the fuel elements which have been exposed to radiation are decomposed and dissolved in the head end of a reprocessing plant, the fission products are released.
Some fission products, such as, for example, krypton and iodine, occur predominantly in the gaseous state and must be removed from the waste gas by means of special devices. However, most fission products are solid and, together with the uranium and the transuraniums generated in the reactor in addition to the fission products, they are dissolved with hot nitric acid during the treatment of the burnt nuclear fuel. After the separation of uranium and plutonium by means of a liquid-liquid extraction, e.g., using as the extraction agent, tributyl-n-phosphate according to the Purex method, the residual aqueous nitrate phase contains practically all of the fission products and the remaining transuranium. In addition, the aqueous solution contains corrosion products and process chemicals from reprocessing.
The amount of waste solution obtained for each kWh of generated electric energy is about 0.004 ml. At first, the waste solutions are stored in steel tanks in the reprocessing plant. Due to their high concentration of activity which, at the beginning, is several thousand curie per liter, they represent a significant potential danger to the biosphere and, therefore, must be carefully isolated. Because of the large amounts of waste solutions generated, in the Federal Republic of Germany, it is projected that there will be an annual generation of about 1560 m.sup.3 by the year 2000, the storage of the highly radioactive wastes in the form of solutions in tanks is not a satisfactory solution of the problem.
For this reason, it is desired to transform the highly radioactive wastes into solids. Solids require less space, they are significantly simpler to handle and store than liquids, and, moreover, they are suited for long-term storage.
However, a solidified product must meet a number of minimum requirements, the most important being:
(a) chemical stability and low solubility in water, PA1 (b) mechanical stability, i.e., the ability to maintain as compact a block shape as possible with a small surface in order to impair dispersion, PA1 (c) good thermal conductivity in order to ensure the removal of the heat from decay and to prevent the build up of high temperatures in the products.
The simplest method of solidification resides in evaporating the waste solution until a solid residue is obtained and decomposing the nitrates by heating to several hundred degrees celsius. This method produces a calcinate which consists predominantly of oxides. However, this calcinate does not meet the three minimum requirements mentioned above.
Therefore, it has been suggested to evaporate the highly radioactive waste solutions, melt the residue and use suitable additives to form glass-like substances. Especially suited as vitrification agents are the oxides of the elements silicon, phosphorus and boron which form a glass network. The combination of silicon dioxide and boron trioxide is usually called borosilicate glass. Phosphate glass is obtained when phosphorus pentoxide is used.
The chemical problems which arise in the glass production are the difficulty of obtaining a homogenous glass with a content of oxides of fission products as high as possible and the corrosion of the melting vessel. As expected, glasses having a high absorptive power for the oxides of fission products, also have a highly corrosive melt. For example, phosphate glass has a high absorptive power and can absorb 30 to 35% by weight of oxides of fission products. In borosilicate glass, the absorptive power is limited to about 20% by weight. In this connection, the element molybdenum is critical due to the formation of a separate phase.
Another problem is the volatilization of highly radioactive fission products during vitrification and especially of the nuclides Cs 137 and Ru 106. Cesium oxide volatilizes to an increasing degree at higher temperatures. The only effective countermeasure is the use of a vitrification temperature as low as possible.
In the waste solution, ruthenium is usually present in the third degree of oxidation. However, in the presence of oxidizing agents and at higher temperatures, it is easily oxidized to the eighth degree of oxidation. RuO.sub.4 is obtained which melts at 25.4.degree. C. and, due to its vapor pressure, infiltrates and significantly contaminates the waste gas.
Since the waste solution contains nitric acid and nitrates, the conditions for oxidation are met. RuO.sub.4 is separated from the waste gas by means of filters containing iron oxide. A procedure frequently used to eliminate the conditions of oxidation is by decomposing (denitration) the nitric acid and nitrates by means of a reducing agent, such as, formaldehyde and formic acid. In this case, alkali nitrates and alkaline earth nitrates are usually not decomposed. However, when phosphoric acid is added, these salts are also subject to dehydration in a displacement reaction. However, the addition of phosphoric acid is practiced only when a phosphate glass is to be subsequently produced.
Thus far, liquid agents have been exclusively used in the denitration stage. Accordingly, the highly radioactive solutions are worked up with formaldehyde solution, sodium nitrate solution and phosphoric acid. Table 1 presents a balance of the solids and the water for a throughput of 200 m.sup.3 of radioactive waste solution, with an operating time of 6000 hours per year:
TABLE 1 ______________________________________ Solids Water (kg/h) (kg/h) ______________________________________ (a) Waste solution (= 200 m.sup.3 /a, = 33.3 l/h = approx. 40 kg/h (approx. 28%) 11.2 28.8 (b) Formaldehyde solution 30% 23.9 l/h = 26.3 kg/h -- 16.7 (c) Reaction water -- 3.6 (d) Sodium nitrate solution 30%, = 15.2 kg/h 1.3 10.6 (e) Orthophosphoric acid 50%, = 10.4 l/h = 14.2 kg/h 7.1 7.1 Total 1 19.6 66.8 Total 2 86.4 (22.7% by weight) Concentration 19.6 29.4 (40.0% by weight) Evaporation -- 37.7 ______________________________________
The evaporated water (37.7 kg/h) represents an unwelcome volume of radioactive secondary waste in the system.